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than a hydride effect; however concentrations of hydrides (rims, blisters) or
low test temperature can overwhelm irradiation effects. Details are given
in the following sub-sections.
4.4.1 Strength and ductility
As outlined in Section 4.3.1, irradiation produces damage in the form of
small dislocation loops (<a> component loops) which harden the material.
The result is an increase in strength and decrease in ductility.
At reactor start-up, the tensile properties are the unirradiated properties
reported by the fuel supplier. Mechanical properties begin changing imme-
diately upon startup, and by an exposure of 5 MWd/KgU or a fl uence of
about 1 × 10 25 n/m 2 ( E > 1 MeV) an increase in strength and decrease in
ductility reach fl uence-saturated values. Figure 4.20 illustrates this point for
Zircaloy-4 irradiated and tested at 315°C (588K) (after Morize et al ., 1987 ).
Note also that the UTSs of cold worked stress relieved (CWSR) and recrys-
tallized (RX) materials become similar at low exposures. This is a general
trend which depends on the balance of hardening by pre-existing disloca-
tions (cold work) and irradiation-produced defects.
Fuel cladding requires suffi cient strength to prevent inward plastic defor-
mation of the cladding at beginning-of-service conditions. PWR strength
must be higher than for BWRs due to the higher water pressure needed to
suppress boiling; therefore, PWR Zircaloy cladding has traditionally been
in the cold work stress relieved annealed (SRA) condition. The discussion
above points out that the difference in strength between SRA and RXA
materials is short-lived under reactor conditions.
700
600
UTS
￿ ￿ ￿ ￿ ￿ ￿
50
500
400
40
CWSR
RX
30
300
20
200
TE
100
10
0
0
0
5
10
15
38
55
100
Fluence, 10E20 n/cm 2
4.20 Effect of neutron fl uence on strength and ductility of recrystallized
(RX) or cold-worked (CWSR) Zircaloy. (Source: Reprinted, with
permission, from Morize et al . (1987), copyright ASTM International,
100 Barr Harbor Drive, West Conshohocken, PA 19428.)
 
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